A conventional boiling water reactor (BWR) includes a pressure vessel containing a reactor core for boiling water to generate steam for powering a steam turbine-generator for generating electrical power, for example. The BWR includes several conventional closed-loop control systems which control various individual operations of the BWR in response to demands.
For example, a conventional recirculation flow control system (RFCS) is used to control core flowrate, which in turn controls output power of the reactor core. A conventional control rod drive system, i.e. rod control system (RCS), controls the control rod position and thereby control rod density within the reactor core for controlling reactivity therein. A conventional feedwater control system controls the feedwater supplied to the pressure vessel, including its flowrate, and thereby the water level within the pressure vessel, and the feedwater temperature is also controlled. And a conventional turbine control controls steam flow from the BWR to the turbine based on lead demands and pressure regulation. All of these systems as well as other conventional systems utilize various monitoring parameters of the BWR for controlling operation thereof. Exemplary conventional monitoring parameters include core flow or flowrate effected by the RFCS, core pressure which is the pressure of the steam discharged from the pressure vessel to the turbine, neutron flux, feedwater temperature and flowrate, steam flow or flowrate provided to the turbine, core power, and various status indications of the BWR systems. Many of the monitoring parameters include conventional monitors or sensors for directly measuring the monitored parameter, whereas other monitoring parameters such as core power are conventionally calculated using other monitoring parameters, and the status monitoring parameters are provided as output signals from the respective systems.
Conventional control parameters which include several of the monitoring parameters listed above are conventionally used for controlling operation of the BWR. The control parameters include, for example, core flow which controls reactor output power, control rod position which controls reactivity in the core, and feedwater flow and temperature which control water level within the pressure vessel and subcooling of the water contained therein, respectively. The several control systems conventionally control operation of the reactor in response to given demand signals such as load demand. A computer program is conventionally used to analyze thermal and hydraulic characteristics of the reactor core for the control thereof, The analysis is based on nuclear data selected from analytical and empirical transient and accident events, and from conventional reactor physics and thermal-hydraulic principles. For example, core response to core flow changes in a BWR is related to conventionally known temperature, Doppler, Void, and power coefficients of reactivity, which reflect the conventional reactor physics and thermalhydraulic principles.
However, in the event of an abnormal transient event the operator on duty in the control room is required to manually react to the event at the very moment of the event based on his training, experience, and judgment. The remedial action taken may or may not be correct depending on the training and knowledge of the operator, and, in the latter event, an unnecessary reactor scram may be required. Furthermore, some transient events may occur exceptionally fast, and faster than the capability of a human operator to react thereto. In such an event, a reactor scram may be automatically effected.
One of the conventional reactor control systems is the nuclear system protection system (NSPS) which is a multi-channel electrical alarm and actuating system which monitors operation of the reactor, and upon sensing an abnormal event initiates action to prevent an unsafe or potentially unsafe condition. The NSPS conventionally provides three functions: (1) reactor trip which shuts down the reactor when certain monitored parameter limits are exceeded; (2) nuclear system isolation which isolates the reactor vessel and all connections penetrating the containment barrier; and (3) engineered safety feature actuation which actuates conventional emergency systems such as cooling systems and residual heat removal systems, for example.
Unless the operator promptly and properly identifies the cause of an abnormal transient event in the operation of the reactor, and promptly effects remedial or mitigating action, the nuclear system protection system will automatically effect reactor trip, which is undesirable if not required.